IN79019
July 17, 1979
MEMORANDUM FOR: B. H. Grier, Director, Region I
J. P. O'Reilly, Director, Region II
J. G. Keppler, Director, Region III
K. V. Seyfrit, Director, Region IV
R. H. Engelken, Director, Region V
FROM: Norman C. Moseley, Director, ROI:IE
SUBJECT: Information Notice No. 79-19, PIPE CRACKS IN STAGNANT
BORATED WATER SYSTEMS AT PWR PLANTS
The subject Information Notice is transmitted for issuance on July 17, 1979.
The Information Notice should be issued to all holders of Reactor Operating
Licenses and Construction Permits. Although a Bulletin action section was
circulated for expedited comment on July 13, 1979 issuance of the Bulletin
is delayed pending accumulation of information by NRR on previous
inspections. Also enclosed is a draft copy of the transmittal letter for
this purpose.
Norman C. Moseley, Director
Division of Reactor Operations
Inspection
Office of Inspection and Enforcement
Enclosures:
1. Information Notice No.
No. 79-19
2. Draft Transmittal Letter
CONTACT: W. J. Collins
49-28190
.
(Draft letter to all power reactors with an operating license or
construction permit)
Information Notice No. 79-19
Addressee:
This Information Notice is provided as an early notification of a possibly
significant matter. It is expected that recipients will review the
information for possible applicability to their facilities. No response is
requested at this time however licensees should be aware that the NRC is
evaluating the issuance of a Bulletin to operating PWR's requesting
information on previous inservice inspections of stagnant borated water
systems and requesting inspection of systems which have not been inspected
recently. If you have questions or comments regarding this matter, please
contact the Director of the appropriate NRC Regional Office.
Sincerely,
Signature
(Regional Director)
Enclosure:
IE Information Notice
No. 79-19
.
UNITED STATES
NUCLEAR REGULATORY COMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON D. C. 20555
July 17, 1979
Information Notice No. 79-19
PIPE CRACKS IN STAGNANT BORATED WATER SYSTEMS AT PWR PLANTS
Description of Circumstances:
During the period of November 1974 to February 1977 a number of cracking
incidents have been experienced in safety-related stainless steel piping
systems and portions of systems which contain oxygenated, stagnant or
essentially stagnant borated water. Metallurgical investigations revealed
these cracks occurred in the weld heat affected zone of 8-inch to 10-inch
type 304 material (schedule 10 and 40), initiating on the piping I.D.
surface and propagating in either an intergranular or transgranular mode
typical of Stress Corrosion Cracking. Analysis indicated the probable
corrodents to be chloride and oxygen contamination in the affected systems.
Plants affected up to this time were Arkansas Nuclear Unit 1, R. E. Ginna,
H.B.Robinson Unit 2, Crystal River Unit 3, San Onofre Unit, 1, and Surry
Units 1 and 2. The NRC issued Circular 76-06 (copy attached) in view of the
apparent generic nature of the problem.
During the refueling outage of Three Mile Island Unit 1 which began in
February of this year, visual inspections disclosed five (5) through-wall
cracks at welds in the spent fuel cooling system piping and one (1) at a
weld in the decay heat removal system. These cracks were found as a result
of local boric acid buildup and later confirmed by liquid penetrant tests.
This initial identification of cracking was reported to the NRC in a
Licensee Event Report (LER) dated May 16, 1979. A preliminary metallurgical
analysis was performed by the licensee on a section of cracked and leaking
weld joint from the spent fuel cooling system. The conclusion of this
analysis was that cracking was due to Intergranular Stress Corrosion
Cracking (IGSCC) originating on the pipe I.D. The cracking was localized to
the heat affected zone where the type 304 stainless steel is sensitized
(precipitated carbides) during welding. In addition to the main through-wall
crack, incipient cracks were observed at several locations in the weld heat
affected zone including the weld root fusion area where a miniscule lack of
fusion had occurred. The stresses responsible for cracking are believed to
be primarily residual welding stresses in as much as the calculated applied
stresses were found to be less than code design limits. There is no
conclusive evidence at this time to identify those aggressive chemical
species which promoted this IGSCC attack. Further analytical efforts in this
area and on other system welds is being pursued.
.
Information Notice No. 79-19 July 17, 1979
Page 2 of 2
Based on the above analysis and visual leaks, the licensee initiated a broad
based ultrasonic examination of potentially affected systems utilizing
special techniques. The systems examined included the spent fuel, decay heat
removal, makeup and purification, and reactor building spray systems which
contain stagnant or intermittently stagnant, oxygenated boric acid
environments. These systems range from 2 1/2-inch (HPCI) to 24-inch (borated
water storage tank suction), are type 304 stainless steel, schedule 160 to
schedule 40 thickness respectively. Results of these examinations were
reported to the NRC on June 30, 1979 as an update to the May 16, 1979 LER.
The ultrasonic inspection as of July 10, 1979 has identified 206 welds out
of 946 inspected having UT indications characteristic of cracking randomly
distributed throughout the aforementioned sizes (24"-14"-12"-10"-8"-2" etc.)
of the above systems. It is important to note that six of the crack
indications were found in 2 1/2-inch diameter pipe of the high pressure
injection lines inside containment. These lines are attached to the main
coolant pipe and are nonisolable from the main coolant system except for
check valves. All of the six cracks were found in two high pressure
injection lines containing stagnated borated water. No cracks were found in
the high pressure injection lines which were occasionally flushed during
makeup operations. The ultrasonic examination is continuing in order to
delineate. the extent of the problem.
Enclosures:
1. IE Circular 76-06
2. List of Information
Notices Issued in 1979
.
November 26, 1976
IE Circular No. 76-06
STRESS CORROSION CRACKS IN STAGNANT, LOW PRESSURE STAINLESS PIPING
CONTAINING BORIC ACID SOLUTION AT PWR's
DESCRIPTION OF CIRCUMSTANCES:
During the period November 7, 1974 to November 1, 1975, several incidents of
through-wall cracking have occurred in the 10-inch, schedule 10 type 304
stainless steel piping of the Reactor Building Spray and Decay Heat Removal
Systems at Arkansas Nuclear Plant No. 1.
On October 7, 1976, Virginia Electric and Power also reported throughwall
cracking in the 10-inch schedule 40 type 304 stainless discharge piping of
the "A" recirculation spray heat exchanger at Surry Unit No. 2. A recent
inspection of Unit 1 Containment Recirculation Spray Piping revealed
cracking similar to Unit 2.
On October 8, 1976, another incident of similar cracking in 8-inch schedule
10 type 304 stainless piping of the Society Injection Pump Suction Line at
the Ginna facility was reported by the licensee.
Information received on the metallurgical analysis conducted to date
indicates that the failures were the result of intergranular stress
corrosion cracking that initiated on the inside of the piping. A commonality
of factors observed associated with the corrosion mechanism were:
1. The cracks were adjacent to and propagated along weld zones of the
thin-walled low pressure piping, not part of the reactor coolant
system.
2. Cracking occurred in piping containing relatively stagnant boric acid
solution not required for normal operating conditions.
3. Analysis of surface products at this time indicate a chloride ion
interaction with oxide formation in the relatively stagnant boric acid
solution as the probable corrodant, with the state of stress probably
due to welding and/or fabrication.
The source of the chloride ion is not definitely known. However, at ANO-1
the chlorides and sulfide level observed in the surface tarnish film near
welds is believed to have been introduced into the piping during testing of
the sodium thiosulfate discharge valves, or valve leakage. Similarly, at
Ginna the chlorides and potential oxygen
.
IE Circular No. 76-06 - 2 - November 26, 1976
availability were assumed to have been present since original construction
of the borated water storage tank which is vented to atmosphere. Corrosion
attack at Surry is attributed to in-leakage of chlorides through
recirculation spray heat exchange tubing, allowing buildup of contaminated
water in an otherwise normally dry spray piping.
ACTION TO BE TAKEN BY LICENSEE:
1. Provide a description of your program for assuring continued integrity
of those safety-related piping systems which are not frequently
flushed, or which contain nonflowing liquids. This program should
include consideration of hydrostatic testing in accordance with ASME
Code Section XI rules (1974 Edition) for all active systems required
for safety injection and containment spray, including their
recirculation modes, from source of water supply up to the second
isolation valve of the primary system. Similar tests should be
considered for other safety-related piping systems.
2. Your program should also consider volumetric examination of a
representative number of circumferential pipe welds by non-destructive
examination techniques. Such examinations should be performed generally
in accordance with Appendix I of Section XI of the ASME Code, except
that the examined area should cover a distance of approximately six (6)
times the pipe wall thickness (but not less than 2 inches and need not
exceed 8 inches) on each side of the weld. Supplementary examination
techniques, such as radiography, should be used where necessary for
evaluation or confirmation of ultrasonic indications resulting from
such examination.
3. A report describing your program and schedule for these inspections
should be submitted within 30 days after receipt of this Circular.
4. The NRC Regional Office should be informed within "24 hours, of any
adverse findings resulting during nondestructive evaluation of the
accessible piping welds identified above.
5. A summary report of the examinations and evaluation of results should
be submitted within 60 days from the date of completion of proposed
testing and examinations.
.
IE Circular No. 76-06 -3- November 26, 1976
This summary report should also include a brief description of plant
conditions, operating procedures or other activities which provide
assurance that the effluent chemistry will maintain low levels of
potential corrodants in such relatively stagnant regions within the
piping.
Your responses should be submitted to the Director of this office, with a
copy to the NRC Office of Inspection and Enforcement, Division of Reactor
Inspection Programs, Washington, D.C. 20555.
Approval of NRC requirements for reports concerning possible generic
problems has been obtained under 44 U.S.C 3152 from the U.S. General
Accounting Office. (GAO Approval B-180255 (R0062), expires 7/31/77)